nrc / epri steam generator task force meetingnrc / epri steam generator task force meeting february...
TRANSCRIPT
NRC / EPRI Steam Generator Task Force
Meeting
February 6, 2013
United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc.
(Indian Point Nuclear Generating Units 2 and 3)
ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 | 05000286 Exhibit #: Identified: Admitted: Withdrawn: Rejected: Stricken:
Other:
NYS000550-00-BD01 11/5/2015
11/5/2015
2 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Agenda
8:30 am Introductions NRC and Industry
Opening Remarks NRC and Industry
EPRI SGMP Steam Generator Task Force Update
• Noise Monitoring
• Anti-Vibration Bar Position Verification
• Divider Plate Cracking
• Tube-to-Tubesheet Weld
• Guidance for Auto Analysis
• Nuclear Safety Advisory Letter 12-01
• Potential Generic Implications of Probe Issue
• Guidance for Performing Inspections Following a Design Basis Event
10:00 am Break
3 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Agenda
10:15 am SGMP Steam Generator Task Force Update (continued)
• Technical Specification Task Force 510
• Upcoming Changes to Industry Documents
• NEI 03-08 Deviations Since February 16, 2012
• ASME Code Changes for Pre-Service Inspection
• Recent Steam Generator Operating Experience
11:00am Response to NRC Comments from August 2012 Meeting
• NRC Comments on ETSS Reports
• Time Dependent Leak Rate
• Industry guidelines pertaining to the use of Nitrogen-16 monitors
11:30 am Address Public Questions/Comments (NRC)
11:45 am Adjourn
4 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Open Technical Issues
5 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Open Technical Issues - SGTF
• Noise Monitoring
– Recommendations have been provided to the Examination
Guidelines Revision 8 Committee
• New appendix has been drafted for noise monitoring
6 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Open Technical Issues - SGTF
• AVB Position Verification
– White paper developed incorporating information from 2012
meeting among SG vendors
– Available to utilities on EPRI website
– Recommendations have been provided to the E&R TAC for
consideration for inclusion in the next revision of the Integrity
Assessment Guidelines
– Westinghouse published NSAL 12-7, “Insufficient Insertion of Anti
Vibration Bars in Alloy 600TT Steam Generators with Quatrefoil
Tube Support Plates”
– SGMP Project began project in 2012 to provide generic input for
plant-specific U-bend tube fatigue analysis
– Follow up survey in 2012 indicates the utilities are familiar with
the issue and are taking actions as appropriate
7 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Open Technical Issues - NRC
• Performance Standards
• EPRI Technical Report 1012984, “Technical Basis for SG
Tube Integrity Performance Acceptance Standards”,
provided to NRC
8 © 2013 Electric Power Research Institute, Inc. All rights reserved.
General Discussion Items
9 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Investigation into Potential Propagation
of Divider Plate Cracking
10 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Divider Plate Crack Propagation
• Roadmap developed to address aging management issue
of Alloy 600 and Alloy 82/182 in steam generator channel
head assembly
– Divider plate crack propagation into channel head
material
– Tube-to-tubesheet weld susceptibility to PWSCC
– UT inspection from outside steam generator
11 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Divider Plate Crack Propagation Roadmap
• Roadmap developed to address
aging management issue of Alloy
600 and Alloy 82/182 in steam
generator channel head
assembly
– Divider plate crack
propagation into channel
head material
– Tube-to-tubesheet weld
susceptibility to PWSCC
– UT inspection from outside
steam generator
12 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Project to Investigate Potential Propagation of
Divider Plate Cracking
• Data and operating experience suggest the divider plate
crack propagation into other areas of the steam generator
channel head is unlikely
– Report was published in June, “Assessment of Channel
Head Susceptibility to PWSCC,” 1025133
• NRC staff provided a copy of this report
• Finite element modeling of the stresses in the channel head
region is complete
• Tube-to-tubesheet weld mockups have been analyzed for
chromium level
– Initial evaluation supports dilution modeling approach
– Report will be published in 2013
13 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Finite Element Modeling of the Channel Head
• 3D FEM of the steam generator channel head was
developed using ANSYS
– Includes bottom channel head, the divider plate, stub
runner, tubesheet, cladding, and associated welds
– Model 51 SG dimensions were used as this was
determined to be the limiting model in Phase 1 of the
project
• A crack is postulated from the divider plate PWSCC
susceptible weld and introduced in the model at the triple
point. The divider plate is assumed to have a 15-inch long
through-wall crack which is considered to be a conservative
assumption
14 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Finite Element Modeling of SG Channel Head
• The specified design basis plant conditions for the Model
51 steam generator is used in the flaw tolerance evaluation
• Thermal transients were developed for the steam generator
hot leg and cold leg side for the design basis plant
conditions
• Results of the FEM stress analyses are post-processed for
use in the fracture mechanics analyses
• Several stress paths are defined through the weld
attachment region between the stub runner, tubesheet, and
channel head wall
– Both hoop and axial stresses are extracted along these paths
– Weld residual stresses are also considered in the analyses
15 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Finite Element Modeling of SG Channel Head
• The allowable flaw depth is determined using ASME Code
Section XI, Appendix A methodology and IWB-3600
acceptance criteria
• Fatigue crack growth evaluation is performed using the
methodology of ASME Code Section XI, Appendix A and
using pc-CRACK software
16 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Conclusions of FEM
• Results show that the maximum stress intensity factor, K, is
less than the allowable values for all design basis plant
conditions up to 75% of the vessel wall thickness.
– Allowable flaw depth is 3.9” into the channel head
• A postulated 0.25” initial circumferential flaw would grow to
a maximum depth of 0.4 inches in 40 years of operation
• A postulated 0.25” initial deep axial flaw would grow about
1 mil in 40 years of operation
– Well below the allowable flaw depth
• The model includes several conservatisms
– An initial flaw in the carbon steel PWSCC resistant
channel head material
– 15” crack in the divider plate
17 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential for Tube to Tubesheet Weld Cracking
• Results of literature review and
data collection reported at the
August 2012 meeting
– Industry has accepted a 24%
chromium content as a
conservative threshold for PWSCC
initiation
– Chromium levels down to 20% have
excellent resistance to initiation
based on testing and operational
experience
• High growth rates during lab tests
are likely due to cold worked test
materials and elevated test
temperatures
• Recent tests indicate that PWSCC
arrests in resistant material
18 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Tube-To-Tubesheet Weld Mockup
82 - Electroslag Welding (ESW)
182 SMAW process
EPRI developed
prototypical tube-
to-tubesheet
mockups to
analyze
chromium levels
19 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Mockup Evaluation
• Mockup from RSG fabricator
with Alloy 82 clad and 690 tube
material
20 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Mockup Weld Geometries Similar
Mockup from RSG Fabricator EPRI Mockup
21 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Area 1
Area 2
Area 3
Area 5
Area 4
Chromium Measurements Taken Confirm the
50% Dilution Algorithm
22 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Conclusions from Analysis of Tube-to-
Tubesheet Weld Mockup
• EPRI mockups shown to represent field based on the RSG
mockup
• Preliminary assessment of the measured chromium content
of autogenous welds validates 50% dilution algorithm
• Chrome recovery suggests high resistance to PWSCC
– 182 weld material tests yield approximately 20% Cr
– 82 weld material tests yield approximately 24% Cr
• Tests will be documented in a technical report in 2013
• Evaluation of stresses in the tubesheet area will be
completed in 2013
23 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Development of a Steam Generator Channel Head
Inspection Technique
• Mockups have been designed
to simulate the SG channel
triple point
– Demonstrate feasibility of
performing UT from the
outside of the SG bowl to
ensure that cracking has not
propagated through the clad
and into the channel head
material
• ASME Code inspection of the
Z-Seam weld cannot be
credited to include the triple
point
24 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Future Work
• Report on SG channel head FEM work will be published in
2013
– Technical work complete
– This report will provide the technical justification for not
inspecting steam generator channel head material
• Report on tube-to-tubesheet weld work will be published in
2013
– Data evaluation ongoing
– Stress analysis of tubesheet region will be performed in
2013
– This report is expected to provide the technical
justification for not inspecting tube-to-tubesheet welds
25 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Guidance for Automatic Data Analysis
26 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Guidance for Automatic Analysis of SG Tube
Eddy Current Inspection Data
• Current industry guidance for automatic data analysis requires two independent data analyses to be performed
• The results are compared and differences resolved
• Industry guidance allows the independent analyses to be manual and/or automatic
• Two utilities, with replacement SGs, have written a technical justification to deviate from the SG Examination Guidelines requirement to perform two independent data analyses
• Both utilities performed single pass automatic analysis on the data from their Alloy 690 SG tubes
• Technical justification for deviation was transmitted to EPRI and NRC in accordance with NEI 03-08
27 © 2013 Electric Power Research Institute, Inc. All rights reserved.
SG Examination Guideline Revision
• A revision of the SG Examination Guidelines is currently ongoing
• As with any guideline revision, the Revision Committee reviews all deviations (since the last revision) and determines if there is a technical basis for changing current guideline requirements
• In addition to the technical justifications written by the utilities that deviated from guideline requirements, other technical information that may be available is also considered
• EPRI has an ongoing project to perform a study that compares the performance of two party manual analysis to single pass automatic data analysis on the same set of field data
28 © 2013 Electric Power Research Institute, Inc. All rights reserved.
EPRI SGMP Project on Single Pass vs Manual
Analysis POD
• The initial phase of the ongoing EPRI project includes analysis of bobbin coil data from a field exam of one SG with Alloy 600MA tubing that was in service for ~20 EFPYs
• The SG was a Model 51 SG and was replaced in 2006
• Degradation mechanisms include ODSCC, PWSCC, wear and thinning at typical locations
• The field primary, secondary and resolution manual data analysis results will be compared to the results from a single pass auto analysis system
• Analysis of the field data with the single pass automatic system is ongoing
• POD information for manual and auto systems will be calculated and binned by eddy current noise
• Initial phase is scheduled to be completed 1st Qtr. 2013
29 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Nuclear Safety Advisory Letter (NSAL)
12-01 Update
30 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Nuclear Safety Advisory Letter (NSAL) 12-01
Update
• A foreign utility reported base metal degradation in the
primary channel head in one of three steam generators
• Unit has been in operation since 1987 and the condition
was observed in Fall 2011
• The visually observed degradation is located only on one
SG, and only on the cold leg side of the channel head in
the vicinity of the bottom bowl drain tube
• Westinghouse communicated to customers via Nuclear
Safety Advisory Letter (NSAL) 12-1
• SGMP communicated to members via e-mail
31 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Cladding Defects
and Wastage of
Channel Head
Base Material
Largest defect in the cladding is 7.7 mm
(0.3 in) x 14.4 mm (0.6 in) by visual
exam
Divider Plate
Divider Plate to
Channel Head WeldChannel Head Cladding
Top of Drain Tube
Flaws in weld and cladding
Not to Scale
32 © 2013 Electric Power Research Institute, Inc. All rights reserved.
EPRI Failure Modes and Effects Analysis
• Joint evaluation effort by EPRI Steam Generator
Management Program, Materials Reliability Program,
Chemistry Program, consultants, and vendor and utility
personnel knowledgeable in materials and boric acid
corrosion
– Identify the most credible failure modes and applicable
degradation mechanisms
– Draft report complete and industry review is ongoing
33 © 2013 Electric Power Research Institute, Inc. All rights reserved.
EPRI Failure Modes and Effects Analysis
Conclusions
• Most likely failure mode is gross defects in stainless steel
cladding that resulted in exposure of low-alloy steel to
concentrated borated water during outages based on the
following assumptions:
– No knowledge of external channel head leakage during
normal operation
• A leak of sufficient magnitude to cause high corrosion
rates would produce ample evidence such that some
form of corrective action would be expected
– Defects present early in the life of the steam generator
permitted concentrated borated water solution in the
channel head region to contact the low-alloy steel
material each outage
• FMEA Report will be published in 2013
34 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Industry Survey
• 44 of 69 units responded to a recent survey regarding SG
channel head configuration and inspections
– 11 have drain lines similar to the drain line described in
the NSAL
• 8 have active drain lines
– Most utilities have inspected or are planning to inspect in
accordance with the recommendations in the NSAL 12-01
regardless of channel head design
• Based on information from the NSAL, some designs
were not considered susceptible to the degradation
• Inspection technique is visual
– No degradation has been reported
35 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential Generic Implications of ARC
Probe Issue
36 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential Generic Implications of ARC Probe
Issue
• Bobbin ARC
– The pulled tube database for bobbin ARC includes data from
both probe types
• Bobbin ETSS(s) used by utilities to size volumetric indications
– The SGMP investigated the ETSSs and found no issue with
regard to the problem of using different bobbin probes and
evaluating their equivalencies.
• Exam Guidelines contain guidance to address the issue
• Bobbin leakage screens in the In Situ Pressure Test Guidelines
– The bobbin voltage screens are for volumetric indications only
– The screening values are conservatively high and would
represent indications that would be of sufficient depth to
require in situ testing for structural integrity
37 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential Generic Implications of ARC Probe
Issue
• NRC Question: Could bobbin probes manufactured from
different suppliers that have variances in voltage amplitude
affect the probability of detection or sizing of tube flaws
– Were the maximum voltage differences considered in
SGMP’s investigation of the generic implications
38 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential Generic Implications of ARC Probe
Issue
• Review of largest voltage delta does not change the
conclusions
• There are no generic implications from this experience
-1.5
-1
-0.5
0
0.5
1
1.5
2
0 500 1000 1500 2000 2500 3000 3500 4000
C17 -C16 voltage delta
Delta
Data point 1138 of 3655 - where Deltas become negative
39 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Potential Generic Implications of ARC Probe
Issue
0
0.5
1
1.5
2
2.5
3
3.5
0 0.5 1 1.5 2 2.5 3 3.5
C
y
c
l
e
1
6
V
o
l
t
a
g
e
s
Cycle 17 voltages
C17vsC16 values
40 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Guidance for Inspections Following a
Design Basis Accident
41 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Guidance for Inspections Following a Design
Basis Accident (DBA)
• Industry Interim Guidance (IG) (SGMP-IG-12-01) was
issued on September 26, 2012 to change the prescriptive
SG inspection requirements contained in the EPRI PWR
SG Examination Guidelines, Revision 7, to performance
based inspection requirements
• The need for IG for forced outages following DBA was
identified following thorough Industry discussions following
a seismic event that exceeded OBE and DBA criteria
42 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Original Prescriptive Guidance in Examination
Guidelines
• The original prescriptive guidance contained in Section
3.10 of the Rev 7 G/L was stated as follows:
“Forced outage examinations shall be performed
during plant shutdown subsequent to any of the
following conditions:
a. SG Primary-to-secondary leakage leading to plant
shutdown
b. Seismic occurrence greater than the Operating
Basis Earthquake
c. Loss-of-coolant accident requiring actuation of the
engineered safeguards
d. Main steam line or feedwater line break”
43 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance
• The IG deleted Section 3.10 of the Rev 7 Exam GL
entirely
• Eliminating the prescriptive inspection requirements
following DBA
• Modified several sections of the SG Integrity Assessment
Guidelines (IAGL) to address performance based
inspections and evaluations following DBA:
44 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance
“Prior to a refueling outage that does not have planned SG
primary-side and/or secondary-side activities, or when the steam
generators have experienced conditions that may not be bounded
by assumptions in the Operational Assessment, the information
used in projecting steam generator integrity in the OA shall be
reviewed. There may have been subsequent plant or industry
experience that impacts the information used in the tube integrity
assessment process that could impact the planned inspection
interval or ability to have an outage without SG inspections.
Following design basis accidents that result in a forced outage,
the review should consider the need for additional analysis,
evaluations, and/or inspections to demonstrate acceptable
condition monitoring and operational assessment prior to start-
up.”
45 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance
“To validate the inspection interval, a review of the DA and OA
information used in evaluating tube integrity shall be performed
prior to each refueling outage when SG primary and/or
secondary-side activities are not scheduled or when the plant
has experienced a forced outage following a design basis
accident. The review should consider the need for additional
analysis, evaluations, and/or inspections to demonstrate
acceptable condition monitoring and operational assessment
prior to start-up.”
46 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance
“Degradation Assessments shall include the secondary
side integrity assessment either by directly incorporating
the results into the DA or by reference. Refueling outages
that do not include secondary side activities (examples:
sludge lancing, FOSAR, secondary side integrity
inspections) or forced outages following design basis
accidents shall be evaluated per section 11.2.4."
47 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance
“A review of the SG integrity assessment documents that justify the
planned inspection interval shall be performed prior to each refueling
outage when SG primary and/or secondary-side activities are not
scheduled, or when the steam generators have experienced conditions
that may not be bounded by assumptions in the Operational
Assessment. The review should consider industry operational
experience, chemistry excursions, plant operating transients, and
design basis accidents since the last inspection and/or review. This
review should be performed in a timely manner in order to minimize
outage impact (scope/budget/planning - critical path), should a change
in the planned inspection interval be identified from the review. For
reviews following design basis accidents that result in a forced outage,
the review should consider the need for additional analysis,
evaluations, and/or inspections to demonstrate acceptable condition
monitoring and operational assessment prior to start-up”.
48 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Basis for Interim Guidance
• Basis for performance based IG for DBA inspections and evaluations:
– Original prescriptive guidance did not specify the inspection scope,
techniques, or criteria and did not tie inspections to SG integrity
requirements
– IG recognizes that each DBA is unique and warrants specific
evaluation of the event to determine the appropriate analyses and
inspections to demonstrate and ensure SG integrity prior to start-up
from the DBA shutdown
– With the uniqueness of each DBA development of specific
prescriptive inspection and evaluation requirements would not be
feasible to capture and address every scenario. This may lead
licensees to miss key aspects of inspection and SG integrity by
relying on prescriptive requirements that may not bound their unique
event.
49 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Other Guidelines for Plant Earthquake
Response
• EPRI Technical Report NP 6695, “Guidelines for Nuclear Plant
Response to an Earthquake.” 1989
• Two NRC Regulatory Guides based on NP 6695
– Regulatory Guides 1.166, “Pre-Earthquake Planning and
Immediate Nuclear Power Plant Operator Post-earthquake
Actions.” 1997
– Regulatory Guides 1.167, “Restart of a Nuclear Power Plant
Shut Down by a Seismic Event.” 1997
• EPRI Report 1025288, Guidelines for Nuclear Plant Response to
Earthquake,” October 2012 updated previous guidance
– Included lessons learned from recent seismic events
• IAGL revision committee will incorporate interim guidance and
consider new technical report in the revision process
50 © 2013 Electric Power Research Institute, Inc. All rights reserved.
TSTF 510 Update
51 © 2013 Electric Power Research Institute, Inc. All rights reserved.
• Information received from 60 of 69 units
Recent Industry Survey Results
TSTF 510
LAR
Approved
Submitted
LAR
(waiting on
approval)
Plan to
Submit
2013
Plan to
Submit
2014
Plan to
Submit later
No Current
Plan to
Submit
Number of
Units 13 13 18 9 3 4*
* 2 units being retired
* 2 units in extended outage
52 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Standing Issues
53 © 2013 Electric Power Research Institute, Inc. All rights reserved.
SGMP Industry Document Status
and Revision Schedule
Guideline Title Current
Rev #
Report # Last
Pub
Date
Imple-
mentation
Date(s)
Interim
Guidance
Review
Date
Comments
SG Integrity
Assessment
Guidelines
3 1019038 Nov
2009
9/1/10 SGMP-IG-
10-01
SGMP-IG-
12-01
2012 Rev 4 in
progress
EPRI SG In Situ
Pressure Test
Guidelines
4 1025132 Oct
2012
10/10/13 None 2015
PWR SG
Examination
Guidelines
7 1013706 Oct
2007
9/1/08 SGMP-IG-
08-04
SGMP-IG-
12-01
Rev 8 in
progress
PWR SG Primary-
to-Secondary
Leakage Guidelines
4 1022832 Sept.
2011
4/11/2012
7/11/2012
None 2014
54 © 2013 Electric Power Research Institute, Inc. All rights reserved.
SGMP Industry Document Status
and Revision Schedule
Guideline Title Current
Rev #
Report # Last
Pub
Date
Imple-
mentation
Date(s)
Interim
Guidance
Review
Date
Comments
PWR Primary
Water Chemistry
Guidelines
6 1014986 Dec
2007
6/17/08
9/17/08
SGMP-IG-
09-01
SGMP-IG-
11-02
Rev 7 in
progress
PWR Secondary
Water Chemistry
Guidelines
7 1016555 Feb
2009
8/20/09
11/20/09
SGMP-IG-
13-01
2013
Steam Generator
Management
Program
Administrative
Procedures
3 1022343 Dec
2010
9/1/11
12/31/11
None N/A
Steam Generator
Degradation
Specific Flaw
Handbook
1 1019037 Dec
2009
N/A None N/A Rev 2 in
progress
55 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance SGMP-IG-13-01
PWR Secondary Water Chemistry Guidelines
• The Interim Guidance modifies existing “shall”
requirements corresponding to when the hydrazine-to-
oxygen ratio is less than two in the water feeding the steam
generators
• This change resulted from a Review Board Inquiry that
identified a gap in guidance for a “loss of feedwater
hydrazine event” (i.e., hydrazine-to-oxygen ratio less than
two in the water feeding the steam generators) that occurs
between low power value (LPV) and mid power value
(MPV) for recirculating steam generators (RSG).
56 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Interim Guidance SGMP-IG-13-01
PWR Secondary Water Chemistry Guidelines
• Expansion: The existing loss of feedwater hydrazine requirement to “commence shutdown” is expanded to mean that the plant must “commence to cold shutdown (RCS ≤ 200°F)”.
• New Guidance: The existing guidance (with the above expansion) on “loss of feedwater hydrazine” for the plant at ≥MPV reactor power for RSGs (≥15% reactor power for OTSGs) is extended to all conditions when the plant is at RCS > 200°F by adding guidance .
• Bases:
– Stress Corrosion Cracking (SCC) is thermally driven
– Hydrazine-to-Oxygen Ratio < 2 causes a significant increase in Electrochemical Potential (ECP), which significantly increases the occurrence and rate of SCC of tubing alloys
57 © 2013 Electric Power Research Institute, Inc. All rights reserved.
NEI 03-08 Deviations
• Four long-term deviations
– Two Steam Generator Examination Guidelines, R7
– Single pass auto analysis
– Steam Generator Secondary Water Chemistry Guidelines, R7
– Wet lay up steam generator sample frequency
– Steam Generator Integrity Assessment Guidelines, R3
– Use of site-specific sizing indices
• Two short term deviations
– Steam Generator Examination Guidelines, R7
– PSI prior to hydro
– Steam Generator Secondary Water Chemistry Guidelines, R7
– Wet lay up steam generator sample frequency
58 © 2013 Electric Power Research Institute, Inc. All rights reserved.
ASME Code Guidance on Pre-Service
Inspection
59 © 2013 Electric Power Research Institute, Inc. All rights reserved.
ASME Code Guidance on Pre-Service Inspection
• NRC Comment regarding Pre-service Examination of
SG tubing:
– IWB-2200(c) of Section XI of the ASME Code indicates
that steam generator tube examination shall be
governed by the plant Technical Specification.
• ASME Section XI Code action BC 10-129 wording: For
Steam Generator tubing a full-length examination on 100%
of the tubing shall be performed using methods and
techniques that are expected to be employed during future
inservice examinations.
60 © 2013 Electric Power Research Institute, Inc. All rights reserved.
ASME Code Guidance on Pre-Service Inspection
• Steam Generator Pre-Service Inspection code requirements are incorporated in an overall steam generator inspection action BC 10-129.
– Meeting held between Code meetings to determine best action to address qualifications and flaws associated with the steam generator inspections.
– Working Group PQSVEC will open a separate Code action to change these qualification requirements.
– Item BC 10-129 is moving forward and will be presented to Subgroup Water Cooled Systems and Subgroup Evaluation and Standards for vote during February 11th Code week.
– ASME Code Section III action being tracked by Subgroup Industry Experience for New Plants.
61 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Operating Experience
62 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant A – ODSCC in A600TT Tubing
• Event Summary: A domestic plant with A600TT tubing
discovered three indications of axial ODSCC in one tube on
the hot leg side during the Fall 2012 steam generator
inspection. One flaw required in situ pressure testing and
passed.
• Pertinent plant operating conditions and design
• Thot 611oF
• 21.27 EFPY
• Stainless steel quatrefoil broach tube support plates
• The affected tube was a high row tube that potentially
contained high residual stress, i.e. -2-sigma tube.
63 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant A – ODSCC in A600TT Tubing
• Three Indications of axial ODSCC found in one tube high row – high
stress tube (-2-Sigma tube)
– TSP 03H (first TSP above flow distribution baffle)
– TSP 05H (second TSP above flow distribution baffle)
– Freespan between TSP 03H-05H (TSP 03H+33.7”)
Indication
+PT Max
Volts
(VPP)
Axial
Length
(inches)
Max Depth
(%TW) Comment
TSP 03H 0.64v 0.58” 69.2% Contained w/in TSP
Land Contact
TSP 05H 0.56v 0.48” 50.0% Contained w/in TSP
Land Contact
Freespan
TSP +33.7” 0.39v 0.19” 56.4%
Coincident with a
1.0v Ding
Appendix I Technique I28432 used for sizing
64 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant A – ODSCC in A600TT Tubing
• The three indications were confirmed by plus-point and the
Ghent probes.
• The indications were not aligned axially, but were ~90
degrees apart.
65 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant A – ODSCC in A600TT Tubing
• All indications were screened for In Situ Pressure Testing
per Rev 3 of the EPRI SG In Situ Pressure Testing
Guidelines
• Indications at TSP 05H and Freespan did not meet criteria
for Proof and Leakage Testing
• Indication at TSP 03H met criteria to perform Proof and
Leakage Testing
– Localized pressure testing was performed up to
pressures associated with SLB and 3xNOP differential
pressure with intermediate pressure tests
– Proof testing and leak testing passed with no leakage
66 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant A – ODSCC in A600TT Tubing
• Condition Monitoring Assessment
– TSP 03H: Met CM limits through In Situ Pressure Testing
– TSP 05H and Freespan: Met CM limits through Flaw Handbook burst pressure requirements (Section 5.1.4)
• Burst Effective (BE) Length/Depth Less than BE Structural Limits
• Leakage criteria met – Max Depth Less than Max Depth Threshold for leakage (i.e., no Pop-Through at SLB conditions)
• Operational Assessment
– All indications met full cycle OA limits through full bundle probabilistic analysis
67 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant B – ODSCC in A600TT Tubing
• Event Summary: A domestic plant with A600TT tubing
discovered four indications of axial ODSCC in two tubes
during the Fall 2012 steam generator inspection. All
indications were on the hot leg side. None of the flaws
required in situ pressure testing.
• Pertinent plant operating conditions and design
– Thot 621oF
– 18.95 EFPY
– Stainless steel quatrefoil broach tube support plates
– The affected tubes did not contain high residual stress
68 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant B – ODSCC in A600TT Tubing
• Three axial indications found in one tube, all located between the flow
distribution baffle (FDB) and the first quatrefoil support plate on the hot
leg side
• One axial indication found in another tube within a hot leg support plate
that contained two dents (11.35v and 8.91v)
Indication
+PT Max
Volts
(VPP)
Axial Length
(inches)
Max Depth
(%TW) Comment
Flaw #1 (FS) 0.96 0.52 77% Coincident with low level
ding
FS Flaw #2 (FS) 0.24 0.15 45% Coincident with low level
ding
FS Flaw #3 (FS) 0.38 0.15 56% Coincident with low level
ding
Flaw #4 (TSP) 0.89 0.22 76% End of Flaw coincident with
the edge of a dent
69 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant B – ODSCC in A600TT Tubing
• One of the freespan indications was initially detected by
bobbin coil and the two smaller indications were found
during the subsequent plus-point diagnostic inspection
– The three indications were not in same axial plane
70 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant B – ODSCC in A600TT Tubing
• The indication at the dented TSP was found during the dent
plus-point inspection program.
– The dent program scope was expanded upon detection
of the axial indication
• Expansion in SG C: 100% dings/dents >5 volts and
20% sample of >2 and <5 volts on HL plus 100%
dings/dents >5 volts and 20% sample of >2 and <5 at
TSP 8C.
• No additional indications found
• All indications were screened for In Situ Pressure Testing
per Rev 3 of the EPRI SG In Situ Pressure Testing
Guidelines
– None of the indications required in situ pressure testing
71 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Domestic Plant B – ODSCC in A600TT Tubing
• Condition Monitoring Assessment
– All indications met CM requirements
• Burst Effective (BE) Length/Depth Less than BE
Structural Limits
• Leakage criteria met – Max Depth Less than Max
Depth Threshold for leakage (i.e., no Pop-Through at
SLB conditions)
• Operational Assessment
– All indications met full cycle OA limits through full bundle
probabilistic analysis
72 © 2013 Electric Power Research Institute, Inc. All rights reserved.
US Operating Experience – Alloy 600TT Tubing
• Majority of the 600TT fleet have H* approved
• Cracking continues to be identified in Alloy 600TT tubing
Location
ODSCC PWSCC
Axial Circ Axial Circ
U-Bend X
TSP/FDB X
TTS/Exp Trans X X X X
Tubesheet X X
Tube End X X
Ding* X
* Some indications at small voltage dings below reporting threshold
73 © 2013 Electric Power Research Institute, Inc. All rights reserved.
US 690TT Units Affected by Wear
MECHANISM
NUMBER OF DOMESTIC UNITS
AFFECTED
Foreign Object Wear 18
U-Bend Support Wear 25
Support Structure Wear 35
Tube-to-Tube Wear 6
Data from Steam Generator Degradation Database as of January 2013
74 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Foreign Plant Tubesheet Leakage Experience
• Tube leakage was detected in 2010 in a tube that was known to have a longitudinal crack within the tubesheet
– 3-loop 900MWe reactor, with AREVA type 51BI steam generators
– 7/8 inch diameter Alloy 600MA tubes (manufactured by Vallourec)
– Tubes were mechanically expanded into the tubesheet by hard rolling and included a kiss roll at the top of the tubesheet
– ~20 steps to achieve mechanical expansion on the full depth of the tubesheet
– The leakage was identified in tube R19C43 with a Helium leak test
• Longitudinal crack = 17mm
17 mm
13 mm
75 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Foreign Plant Tubesheet Leakage Experience
• A root cause evaluation has shown that the
tube was manufactured within the
tolerances; however a stack up of
tolerances could explain the leakage
• Verified that the indication was PWSCC
76 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Response to NRC Comments from
August 2012 Meeting
77 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Eddy Current Essential Variable
Tolerance
78 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Eddy Current Essential Variable Tolerance
• The EPRI Examination Technique Equivalency Project was
planned as a multi-phase, multi-year project to provide a
consistent and cost-effective methodology to allow utilities
to determine equivalency of eddy current techniques that
differed in one or more essential variables from EPRI-
qualified techniques
•Three EPRI reports have been issued
o Development of a Process for Determining Examination Technique
Equivalency, (1015126), March 2008
o Development of Standardized Process for Determining Examination
Technique Equivalency, (1018557), March 2009
o Development of Documentation for Examination Technique
Equivalencies, (1020992), September 2010
79 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Eddy Current Essential Variable Tolerance
• The first report documented a method for demonstrating
equivalency between EPRI-qualified eddy current inspection
techniques for SG tubes and similar inspection techniques
with modified essential variables.
– A master set of calibration tubes was designed.
• The second report documented a review of various EPRI-
qualified techniques compared to field techniques with
modified essential variables.
– Tolerances for technique equivalency were established.
• The third report provided information for developing
validation documentation for site inspection techniques with
modified essential variables. This phase of the project further
documented the effect of changes to essential variables on
the resulting eddy current signal.
80 © 2013 Electric Power Research Institute, Inc. All rights reserved.
NRC Review of EPRI Examination Technique
Equivalency Reports
• April 19, 2012 - All 3 reports were made available for NRC
review at EPRI Charlotte office
• August 21, 2012 - NRC provided comments at SGTF
meeting
• October 24, 2012 - EPRI met with NRC contractor to review
comments
81 © 2013 Electric Power Research Institute, Inc. All rights reserved.
EPRI Plans Going Forward
• Request continued interaction with NRC NDE contractor to
receive clarification of initial comments
• Propose a 2014 project to issue a fourth report that:
– Addresses equivalency topics not addressed in previous
reports
– Considers comments received from utility members,
utility contractors and NRC
82 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Time Dependent Leak Rates
83 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Time Dependent Leak Rates
• EPRI Technical Report 1022831, “Onset of Fatigue Cracking
in Steam Generator Tubes with Through Wall Flaws”,
provided to NRC
• NRC question: The conclusions in this report are based
on room temperature tests. Would similar conclusions
be made if the tests were at operating temperatures?
– Increase in time dependent plasticity at higher
temperatures because of higher creep rate?
– Increase in jet-structure interaction at higher
temperatures
• The following slides describe recent test results
84 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Time Dependent Deformation of Alloy
600
• Time dependent deformation can lead to time dependent
increases in leak rates of degraded tubing
• Previous testing indicated small levels of time dependent
deformation in Alloy 600 at room temperature
• Deformation levels of practical interest were found to be
essentially complete at 300 seconds
• Time dependent deformation is approximately equivalent to
a 2% decrease in material flow strength
• No adjustments to in situ testing practice required given
the conservative test procedure
• Testing recently performed at 600°F to evaluate the
possibility of increased levels of time dependent
deformation
85 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Stress Strain Curves for 600MA Material at
Room Temperature and 600°F
• Typical Decrease in Strength is Observed
Mill Annealed Alloy 600
0
10000
20000
30000
40000
50000
60000
70000
80000
90000
100000
0 0.05 0.1 0.15 0.2 0.25 0.3 0.35
Engineering Strain, in./in.
En
gin
ee
rin
g S
tre
ss
, p
si
Room Temperature
600° F
86 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Step Load Test at Room Temperature
• Small Levels of Time Dependent Deformation are Observed
0
0.05
0.1
0.15
0.2
0.25
0 2000 4000 6000 8000 10000 12000 14000 16000 18000
Time, seconds
En
gin
eeri
ng
Str
ain
, in
./in
.
45.4 ksi
51.9 ksi
58.3 ksi
64.7 ksi
71.1 ksi
77.5 ksi
83.9 ksi
87 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Magnified View of Time Dependent Deformation
at Room Temperature
• Deformation is Essentially Complete at 300 Seconds
0
0.005
0.01
0.015
0.02
0.025
0.03
0.035
0 50 100 150 200 250 300 350
Time, seconds
Incre
ase in
Str
ain
, in
./in
.
45.4 ksi
77.5 ksi
88 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Step Load Test at 600°F
• Small Levels of Time Dependent Deformation are Observed
0
0.05
0.1
0.15
0.2
0.25
0 300 600 900 1200 1500 1800 2100 2400
Time, seconds
En
gin
eeri
ng
Str
ain
, in
./in
.
600 °F
45.4 ksi
51.9 ksi
58.3 ksi
64.9 ksi
71.1 ksi
77.5 ksi
89 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Magnified View of Time Dependent Deformation
at 600°F
• Deformation is Essentially Complete at 300 Seconds
0
0.005
0.01
0.015
0.02
0.025
0.03
0.035
0 50 100 150 200 250 300 350
Time, seconds
Incre
ase i
n S
train
, in
./in
.
600 °F
45.4 ksi
77.5 ksi
90 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Conclusions
• Time Dependent Deformation at a Level of Practical
Interest Relative to Time Dependent Leak Rates is the
Same at 600 °F as It is at Room Temperature
• No adjustments to in situ test practice are required
• No adjustments to leak rate calculations are required
91 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Industry Guidelines Pertaining to the Use
of Nitrogen-16 Monitors
92 © 2013 Electric Power Research Institute, Inc. All rights reserved.
NRC Question
• Do guidelines address the use (or lack of use) of
confirmatory measurements when a rapidly increasing leak
rate is only detected on the faster response time monitor?
93 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Primary to Secondary Leak Guidelines Requirements
• Guidelines allow qualitative confirmation to prevent unnecessary
shutdowns
– N-16 monitors installed on or near steam lines are very good at
quickly detecting tube primary-to-secondary leakage, but may not
provide an accurate leak rate value.
– N-16 monitor indications may be an anomalous indication from an
instrument issue or EMF interference, etc. which has been
experienced in the industry
• Guidelines (Action Level 3) require plants without a second radiation
monitor shut down based on the indication of one monitor
• Guidelines point out response time of radiation monitors
– Plant personnel evaluate response times and reflect it in plant
procedures and training programs so that responses occur within
appropriate time frames
94 © 2013 Electric Power Research Institute, Inc. All rights reserved.
NRC Discussions/Items of Interest
95 © 2013 Electric Power Research Institute, Inc. All rights reserved.
Together…Shaping the Future of Electricity